HomeMy WebLinkAbout851174.tiff SSINS NO. : 6835
IN 85-72
UNITED STATES WELD cp ,i,y
NUCLEAR REGULATORY COMMISSION wf
OFFICE OF INSPECTION AND ENFORCEM ' , •,_
WASHINGTON, D.C. 20555o
August 22, 1985 Adlp0 91985 ir
GREEIEY, Cp
La
IE INFORMATION NOTICE NO. 85-72: UNCONTROLLED LEAKAGE OF REACTOR COOLANT
OUTSIDE CONTAINMENT
Addressees:
All boiling water reactors holding an operating license (OL) or a construction
permit (CP).
Purpose:
This information notice is provided to alert recipients of a significant event
involving an uncontrolled primary coolant leak outside containment. It is
expected that recipients will review the information for applicability to
their facilities and consider actions, if appropriate, to preclude similar
events from occurring at their facilities. However, suggestions contained in
this information notice do not constitute requirements; therefore, no specific
action or written response is required.
Description of Circumstances:
On June 12, 1985, a reactor scram occurred from 99% power at Oyster Creek
Nuclear Generating Station. The scram occurred following failure of the elec-
tric pressure regulator that subsequently caused a turbine bypass valve to
open. This resulted in a reactor pressure decrease to the low pressure trip
set point, causing the main steam isolation valves (MSIVs) to shut and the
reactor to scram.
As part of the scram sequence, the scram discharge volume (SDV) vent and drain
valves are required to shut to contain the water released during a scram.
However, in this event, the two drain valves did not fully close, allowing
hot, reactor coolant to drain to the reactor building equipment drain tank.
The hot fluid flashed in the drain system creating steam that flowed up through
various drains in the 51-ft and 23-ft building levels. The steam combined with
the fumes from the blistering paint on the SDV drain piping and caused a por-
tion of the reactor building deluge fire protection system to actuate and spray
down the 51-ft level . Approximately 500 gal of reactor coolant flowed to the
drain tank before the scram system could be reset, which took approximately 38
minutes. The fire protection deluge system actuated approximately 20 minutes
after the scram and was shut off in approximately 5 minutes. No safety equip-
ment inside the reactor building was adversely affected by actuation of the
deluge system.
8508200630
851174
IN 85-72
August 22, 1985
Page 2 of 3
Discussion:
The failure of the SDV drain valves to properly close caused the following:
1. Uncontrolled reactor coolant leakage outside containment.
2. Temperatures of the control rod drive (CRD) seals exceeding the alarm
setting.
3. Actuation of the reactor building fire protection deluge system.
4. Radioactive contamination of the 23-ft level of the reactor building.
Each SDV drain valve that failed had a different failure mechanism. The
upstream valve stem/disc travel stopped approximately 1/8 inch before fully
seating onto the valve seat. This was caused by the valve actuator not having
the stroking length properly adjusted. The downstream valve had an improperly
sized spring in the valve actuator. It is believed that the valve initially
closed, but was then forced open when the system pressure exerted a force
below the valve seat that exceeded the spring closing force of the actuator.
The high CRD seal temperature alarms were received intermittently after the
scram. The alarms are an indication of abnormal flow of reactor coolant within
or out of the CRD system. Degradation, or possibly failure, of the seals could
occur following prolonged exposure at elevated temperatures. As a result,
abnormal leakage could occur that might adversely affect proper rod motion or
rod scramming ability.
The reactor building fire protection system actuated on the 51-ft level of the
reactor building. Although no equipment was adversely affected by the deluge
system spray the potential existed for damaging electrical equipment and pos-
sibly aggravating an already serious problem.
Although the SDV vent and drain valves are stroke tested monthly in accordance
with the inservice testing (IST) program, there were no criteria or requirements,
for leak testing these valves. Following the initial installation of the down-
stream valve as part of a system backfit in 1984, no postinstallation leak rate
test of either valve against operating pressure was conducted. Both valve
problems could have been detected by such a test.
IE Information Notice 84-35, "BWR Post-Scram Drywell Pressurization" described
an event of August 1982 at the Hatch Nuclear Plant Unit 2 where there was a
similar leakage from the SDV. That event was also the subject of an AEOD case
study and was included in the 3rd quarter, 1983, "Report to Congress on
Abnormal Occurrences."
IN 85-72
August 22, 1985
Page 3 of 3
No specific action or written response is required by this information notice.
If you have any questions regarding this matter, please contact the Regional
Administrator of the appropriate NRC regional office or this office.
1
764/11144--
d Lrdan, Director
Divisi of Emergency Preparedness
. and ngineeringg Response
Office of Inspection and Enforcement
Technical Contact: David Powell , IE
(301) 492-8373
Attachment: List of Recently Issued Information Notices
Attachment 1
IN 85-72
August 22, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
information Date of
Notice No. Subject Issue Issued to
85-71 Containment Integrated Leak 8/22/85 All power reactor
Rate Tests facilities holding
an OL or CP
85-70 Teletherapy Unit Full 8/15/85 All material
Calibration And Qualified licensees
Expert Requirements (10 CFR
35.23 And 10 CFR 35.24)
85-69 Recent Felony Conviction For 8/15/85 All power reactor
Cheating On Reactor Operator facilities holding
Requalification Tests an OL or CP
85-68 Diesel Generator Failure At 8/14/85 All power reactor
Calvert Cliffs Nuclear facilities holding
Station Unit 1 an OL or CP
85-42 Loose Phosphor In Panasonic 8/12/85 Materials and fuel
Rev. 1 800 Series Badge Thermo- cycle licensees
luminescent Dosimeter (TLD)
Elements
85-67 Valve-Shaft-To-Actuator Key 8/8/85 All power reactor
May Fall Out Of Place When facilities holding
Mounted Below Horizontal Axis an OL or CP
85-66 Discrepancies Between 8/7/85 All power reactor
As-Built Construction facilities holding
Drawings And Equipment an OL or CP
Installations
85-65 Crack Growth In Steam 7/31/85 All PWR facilities
Generator Girth Welds holding an OL or CP
85-64 BBC Brown Boveri Low-Voltage 7/26/85 All power reactor
K-Line Circuit Breakers, With facilities holding
Deficient Overcurrent Trip an OL or CP
Devices Models OD-4 and 5
OL = Operating License
CP = Construction Permit
SSNSINS No. : 6835
WELD tuuN8IP5 C7
1
UNITED STATES D F1
NUCLEAR REGULATORY COMMISSION I EI VE
OFFICE OF IOD.C. 0555
AND 20555
AUG 3 9 1985
WASHINGTON,
August 22, 1985 GREELEY, COLO.
IE INFORMATION NOTICE NO. 85-71: CONTAINMENT INTEGRATED LEAK RATE TESTS
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP):
Purpose:
This information notice is provided as a notification of a potentially signifi-
cant problem pertaining to containment integrated leak rate tests (CILRTs). It
is expected that recipients will review the information for applicability to
their facilities and consider actions, if appropriate, to preclude a similar
problem occurring at their facilities. However, the suggestion contained in
this information notice (namely, that licensees review their programs with
respect to the guidelines provided), does not constitute an NRC requirement.
Therefore, no specific action or written response is required.
Description of Circumstances:
Recent staff reviews of the CILRTs performed at San Onofre, Kewaunee, and
Monticello nuclear power plants have indicated that many utilities are misin-
terpreting the relationship between local leak rate testing and CILRTs. 10 CFR
50, Appendix J, discusses containment leakage testing in terms of Type A, B,
and C tests. The Type A test is a measurement of the overall integrated
leakage rate of the primary containment; whereas Type B and C tests are local
leak rate tests designed to detect and measure local leakage across each
pressure-containing or leakage-limiting boundary for primary containment.
As a result of Type B and C tests, some utilities are performing repairs and
adjustments before conducting Type A tests without properly adjusting the Type
A test results for the Type B anti C leakage rates. Without this adjustment,
the "as found" condition of the primary containment cannot be properly determined.
In some cases, when this adjustment is made properly, a Type A test may fail to
meet the acceptance criteria of Apprndix J with regards to the "as found"
condition. When two successive Type A test failures occur, Appendix J requires
more frequent CILRTs. However, if Type B and C leakage rates constitute an
identified contributor to this failure of the "as found" condition for the
CILRT, the general purpose of maintaining a high degree of containment integrity
might be better served through an improved maintenance and testing program for
8508200623
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IN 85-71
August 22, 1985
Page 2 of 4
containment penetration boundaries and isolation valves. In this situation,
the licensee may submit a Corrective Action Plan with an alternative leakage
test program proposal as an exemption request for NRC staff review. If this
submittal is approved by the NRC staff, the licensee may implement the corrective
action and alternative leakage test program in lieu of the required increase in
Type A test frequency incurred after the failure of two successive Type A
tests.
Discussion:
Sections III.D.1.a, 2.a and 3.a of Appendix J require that a set of three
periodic Type A tests be performed at approximately equal intervals during each
10-year service period, and that Type B and C tests be performed during reactor
shutdown for refueling but in no case at intervals greater than 2 years. Under
these test requirements, there are many occasions when Type A, B, and C tests
must be performed during the same reactor shutdown period. Questions are
frequently raised concerning the correct sequence of conducting the Type A, B,
and C tests and the potential impact of the results of the Type B and C tests
on the success or failure of a periodic Type A test.
The NRC staff has previously provided partial guidance to utilities on these
questions on an individual case basis with respect to inspection and enforce-
ment activities (see Attachment 1). The staff position on these questions, as
previously employed in inspection and enforcement, may be summarized as
follows:
1. Section III.A.3 of Appendix J requires that all CILRTs be conducted in
accordance with the provisions of ANSI N45.4-1972. Paragraph 4.2 of ANSI
N45.4-1972 states that for periodic CILRTs no repairs or adjustments are
to be made to the containment structure prior to conducting the test in
order to disclose the normal state of repair of the containment structure.
2. Type B and C tests may be performed either before the start of or after
completion of the periodic Type A test provided that the pretest require-
ments of Paragraph 4.2 of ANSI N45.4-1972 and Section III.A.1.a of Appen-
dix J are met; i.e. , no repairs or adjustments to the primary containment
boundary are made so that the containment can be tested in as close to the
"as is" condition as practical . As such, the leakage information obtained
from the "as is" (sometimes called "as found") Type A test results can be
used to assess the containment condition and its integrity following a
period of plant operation.
3. If repairs or adjustments performed as a result of the Type B and C
testing programs or for any other reasons are made to the primary contain-
ment boundary before the Type A test sequence, local leak tests must be
performed on the affected portion of the containment boundary to determine
the minimum pathway leakage rates before and after the repairs or adjust-
ments are made. The minimum pathway leakage would be the smaller leakage
rate of in-series valves tested individually, one-half the leakage rate
IN 85-71
August 22, 1985
Page 3 of 4
for in-series valves tested simultaneously by pressurizing between the
valves, and the combined leakage rate for valves tested in parallel . The
"as found" Type A test results can then be obtained by adding the differ-
ences between the affected minimum path leakage before and after repairs
or adjustments to the overall measured Type A test result. A periodic
Type A test would be called a "failure" if the "as found" Type A test
result (with appropriate correction from local leak tests) exceeds the
acceptance crtieria of Appendix J.
4. The question has been raised by various utilities as to how far in advance
of the Type A test the Type B and C tests may be conducted without having
to add the leakage differences to the Type A test results. The staff
position on this question has been that after Type B and C tests, the
penetrations and valves should experience some period of normal service
conditions before the Type A test. If the Type B and C tests are con-
ducted before the Type A test during the same refueling outage, the
service condition criterion would not be met. If, however, some operating
service time is achieved, the Type A test can be conducted essentially
independent of the time duration of exposure to the normal service condi-
tions. Thereafter, a Type A test could be conducted without having to
consider the local leak rate results in determining the "as found"
condition.
The continuance of containment leak-tight integrity is the primary importance
in performing Type A, B, and C tests. Therefore, it may be beneficial for li-
censees to implement improved maintenance and testing programs for containment
penetrations to ensure that known or likely leaking penetrations will not result
in the overall loss of containment leak-tight integrity and in the ensuing
penalties for Type A test failure.
It should also be noted that containment leak-tight integrity is monitored
between CILRTs through the Type B and C test programs. Failure to meet the
acceptance criteria of Appendix J for those tests generally constitutes a loss
of containment integrity as defined in the Technical Specifications and may be
reportable by the licensee under the provisions of 10 CFR 50.73. , Sections
(a)(2)(ii) and (a)(2)(v)(C).
It is suggested that licensees review their CILRT program with respect to the
above guidelines.
IN 85-71
August 22, 1985
Page 4 of 4
No specific action or written response is required by this information notice;
however, if you have any questions regarding this notice, please contact the
Regional Administrator of the appropriate NRC regional office or the technical
contacts listed below.
'Edward 1. rdan�tor
Division o Emergency Preparedness
and En. neering Response
Office of Inspection and Enforcement
Technical Contacts: Y. S. Huang, NRR
(301) 492-9493
D. C. Kirkpatrick, IE
(301) 492-4510
S. A. McNeil , IE
(301) 492-9602
Attachments:
1. Documentation from NRC to Utilities, Related to Repairs and
Adjustments Done Prior to Type A Tests
2. List of Recently Issued IE Information Notices
Attachment 1
IN 85-71
August 22, 1985
Page 1 of 2
Documentation from NRC to Utilities, Related to
Repairs and Adjustments Done
Prior to Type A Tests
1. Letter to Consumers Power Company from R. L. Spessard, "Big Rock Point
CILRT Schedule," February 3, 1983
This letter informed the licensee of the necessity to increase the CILRT
frequency because of the failure of two consecutive Type A tests conducted
in 1977 and 1982. During the 1982 refueling outage, Type B and C tests
were conducted and several valves were found to leak excessively and were
repaired. Subsequently the Type A test was conducted and the licensee
reported a successful test, but it did not include the initial Types B and
C leakage in the Type A test results. The NRC staff reviewed the tests
and determined that the Type B and C leakage should be added to the Type
, A test results, because the plant had not been in service between the time
of the Type B and C tests and the Type A test. With the addition of the
Type B and C leakage to the Type A test result, the leakage was excessive
and the containment was deemed to have failed the "as found" test
condition.
2. Letter to Commonwealth Edison Company from R. L. Spessard, "Quad Cities
Unit 1 Containment Integrated Leak Rate Test Frequency," October 7, 1983.
This letter also informed the licensee of the necessity to increase the
CILRT frequency because of the failure of two consecutive Type A tests.
These tests were conducted in 1979 and 1982. Type B and C tests conduct-
ed during the 1982 refueling outage, prior to the Type A test, showed that
the combined leakage from several valves exceeded the allowable Technical
Specification. In addition, the seal between the drywell head and the
drywell vessel flange was found to be leaking to such an extent that the
leakage could not be measured. The licensee repaired these leaks and then
conducted a Type A test that showed the leakage to be within the allowable
limits. The NRC staff, however, determined that the containment had
failed the CILRT with respect to the "as found" condition. This determi-
nation was based on the position that the Type B and C test results could
be excluded from the "as found condition" only if some period of normal
station service existed between Type B and C tests and the Type A test.
3. Inspection Report No. 50-305/84-19 (DRS), Kewaunee, November 27, 1984 and
Notice of Violation to Wisconsin Public Service Corporation - Docket No.
50-305, November 28, 1984.
Attachment 1
IN 85-71
August 22, 1985
Page 2 of 2
The inspection report discusses an exemption to Appendix J issued to
Wisconsin Public Service Corporation by the NRC. The exemption permitted
Type B and C tests and repair work on penetrations to be performed at
Kewaunee before Type A tests were conducted. The exemption required that
leakage reduction caused by the repairs be added to the Type A test result
for the purpose of evaluating the "as found" condition. The licensee then
wrote to the NRC stating that it did not believe that an exemption was
required to perform Types B and C tests before performing a Type A test.
The licensee based this on the belief that Type A testing and Type B and
C testing were two separate events performed on two separate schedules.
In 1984, the licensee performed Type B and C tests before performing the
Type A test and failed to add the pre- and post-repair differential
leakage to the "as found" Type A test results in its CILRT report. As
stated in the inspection report, the NRC staff did not agree with the
licensee' s position because Type B and C testing (with repair) would
invalidate part of the purpose of the Type A test (that is, to establish
the "as found" condition). As a result, the notice of violation covering
this failure was issued on November 28, 1984.
4. Inspection Report No. 50-206/85-12 San Onofre Unit 1, April 5, 1985.
Paragraph 6 of this report discusses the results of the CILRT performed at
San 0nofre during1985. Type C testing and repair work was performed on
six sets of valves just before the Type A test was conducted. However,
differential leakage resulting from the repair was not added to the Type A
test results reported. As a result a notice of violation covering this
failure is under consideration.
Attachment 2
IN 85-71
August 22, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-70 Teletherapy Unit Full 8/15/85 All material
Calibration And Qualified licensees
Expert Requirements (10 CFR
35.23 And 10 CFR 35.24).
85-69 Recent Felony Conviction For 8/15/85 All power reactor
Cheating On Reactor Operator facilities holding
Requalification Tests an OL or CP
85-68 Diesel Generator Failure At 8/14/85 All power reactor
Calvert Cliffs Nuclear facilities holding
Station Unit 1 an OL or CP
85-42 Loose Phosphor In Panasonic 8/12/85 Materials and fuel
Rev. 1 800 Series Badge Thermo- cycle licensees
luminescent Dosimeter (TLD)
Elements
85-67 Valve-Shaft-To-Actuator Key 8/8/85 All power reactor
May Fall Out Of Place When facilities holding
Mounted Below Horizontal Axis an OL or CP
85-66 Discrepancies Between 8/7/85 All power reactor
As-Built Construction facilities holding
Drawings And Equipment an OL or CP
Installations
85-65 Crack Growth In Steam 7/31/85 All PWR facilities
Generator Girth Welds holding an OL or CP
85-64 BBC Brown Boveri Low-Voltage 7/26/85 All power reactor
K-Line Circuit Breakers, With facilities holding
Deficient Overcurrerit Trip an OL or CP
Devices Models OD-4 and 5
85-63 Potential for Common-Mode 7/25/85 All power reactor
Failure of Standby Gas Treat- facilities holding
ment System on Loss of Off- an OL or CP
Site Power
OL = Operating License
CP = Construction Permit
- o°EPa aec1/4gJ UNITED STATES
°
. NUCLEAR REGULATORY COMMISSION
I ) '' I ;
m ; REGION IV
1' 611 RYAN PLAZA DRIVE, SUITE 1000
do -, ° ARLINGTON,TEXAS 76011 h'(!P p;,,.,,,,.
AUG 2 7 1985 (
('
Docket: 50-267 I( `r ,
AUG 2 g 1985
GREELEY.
Public Service Company of Colorado coLo.
ATTN: 0. R. Lee, Vice President
Electric Production
P. 0. Box 840
Denver, Colorado 80201-0840
Gentlemen:
We have reviewed your December 30, 1983, application "Proposed Changes to the
Inservice Inspection and Testing Requirements" and find that additional
information is needed. The findings of our review were discussed with members
of your staff during meetings held during the week of July 22, 1985, on the
Technical Specifications Upgrade Program.
We have coordinated our review with corresponding portions of the Fort St.
Vrain Technical Specification Upgrade Program now in progress. This has been
done by associating as well as possible the corresponding "Draft Items" from
the upgrade program with each of the surveillances reviewed. In doing this we
have found many Draft Upgrade Items which are unsatisfactory, although in a
few cases we find them superior to the proposed surveillances. Identification
and discussion of these items now appears in the enclosure (identified by
following the assigned number with a "U" for upgrade).
Major items of importance are:
The December 30, 1983, submittal does not complete the surveillance
program PSC committed to implement. We observe that, together with the
material addressed by Fort St. Vrain License Amendment 33, the current
review deals with the higher priority items but some Category II systems
remain as well as the majority of the Category III and IV systems.
Whether these surveillances should be addressed in the context of the
present technical specifications or as part of the upgrade program should
be based on the anticipated rate of progress of the upgrade program.
° Many of the systems we addressed were reviewed in the context of the
July 1, 1983, edition of the ASME Code which includes "Rules for
Inspection and Testing of Components of Gas-Cooled Plants ," (Section XI,
Division 2). This reference expedited our review, provided suitable
comparability with light water reactors, and otherwise strengthened the
quality of the Fort St. Vrain surveillance program. It is our
recommendation that this Code be used wherever practicable. While there
may be some necessary exceptions both parties will have to take, our
review has illustrated the benefits of consistency, completeness and
efficiency in spite of some potential difficulties. This practice would
be similar to that used for older LWRs.
9.
I S/Hlz,
Public Service Company -2-
of Colorado
0 Fort St. Vrain License Amendment 33 and document prepared by ASIA Inc. ,
which formed part of the basis were also reviewed. One of the ASTA
recommendations was that PSC investigate the application of visual
examination techniques to certain portions of the thermal barriers in the
lower plenum. While it was concluded in the SER supporting Amendment 33
that no practical access was available for such examinations, we
believe that this issue should not remain closed. It is our opinion that
PSC should commit to a continuing awareness of advancements in inspection
techniques, particularly those involving miniaturization of
instrumentation, and be prepared to implement such inspections should the
technology become available.
A copy of our evaluation is enclosed for your review and comment. It is our
understanding, based on discussions at the above mentioned meeting, that PSC
will provide a resubmittal of these requirements within 90 days of your
receipt of this evaluation.
Since this reporting requirement relates solely to the Fort St. Vrain Station,
OMB clearance is not required under P.L. 96-511.
Sincerely,
Dorwin R. Hunter, Chief
Reactor Safety Branch
Enclosure:
Evolution of ISI/TSI
cc:
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. 0. Box 85608
San Diego, California 92138
Kelley, Stansfield & O'Donnell
Public Service Company Building
550 15th Street, Room 900
Denver, Colorado 80202
Public Service Company -3-
of Colorado
/Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Manager
Nuclear Licensing/Fuels Div.
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 19}
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
Colorado Radiation Control Program Director
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